ML19301A162

From kanterella
Jump to navigation Jump to search
Auxiliary Feedwater Sys Automatic Initiation & Flow Indication,Vepco,North Anna Unit 1, Technical Evaluation Rept
ML19301A162
Person / Time
Site: North Anna Dominion icon.png
Issue date: 11/03/1981
From: Fertner K, Vosbury F
FRANKLIN INSTITUTE
To: Kendall R
NRC
Shared Package
ML19301A163 List:
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-288, NUDOCS 8111060016
Download: ML19301A162 (15)


Text

. - _ _ _ _

l TECHNICAL EVALUATION REPORT AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION VIRGINIA ELECTRIC AND POWER COMFANY NORTH ANNA UNIT 1 f.RC DOCKET NO. 50-338 FRC PROJECT CS257 NRCTACNO. 43191 FRC ASSIGNMENT' 9 NRC CONTRACT NO. NRC-03-79-118 FRC TASK 288 Preparedby Authot

  • F. Vosbury Franklin Research Center The Parkway atTwentieth Street Philadelphia, PA 19103 FRC Group Leader: K. Fertner Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: R. Kendall November 3, 1981 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or respont lbility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this repott, or represents that its use by such third party would not infringe privately owned rights.

4 Franklin Research Center A Division of The Franklin institute e111060016 CF ADOCMof00 g CF

TECHNICAL EVALUATION REPORT AUXILIARY FEEDWATER SYSTEM AUTOMATIC INITIATION AND FLOW INDICATION

VIRGINIA ELECTRIC AND POWER COMPANY l NORTH ANNA UNIT 1 mmmmmmmmmmmmma i NRC DOCKET NO. 50-338 FRC PROJECT C5257 NRC TAC NO. 4 3191 FRC ASSIGNMENT 9 NRC CONTRACT NO NRC-03-79-118 FRC TASK 288 Prepared by Franklin Research Center Author
F. Vosbury The Parkway at Twentieth Street i

Philadelphia, PA 19103 FRC Group Leader: K. Fertner Preparedfor '

Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: R. Kendall ,

1 November 3, 1981 This report was prepared as an account of work sponsored by an agency of the United States Gowrnmert. Neither the United Ststes Government nor any agency thereof, or any of their employees, i makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe orivately owned rights.

l l

A 1 i

_ L _ Franklin Research Center A Division of The Franklin institute .

The Benjamsn Frankhn Parkway Phda.. Pa. 19103(215)448-1000

,. r - - - - ,

.3 yy- 4

TER-C5257-288

,' o l CONTENTS l

Section ' Title Page

)

i 1 INTRODUCTION . . . . . . . . . . . . . .1 i 1.1 Purpose.of Review . . . . . . . . . . 1 I t

l 1.2 Generic Issue Backgrocnd . . . . . . . . 1  !

1.3 Plant-Specific Background . . . . . . . . 2 j 2 REVIEW CRITERIA . . . . . . . . . . . . 3 ,

3 TECHNICAL EVALUATION . . . . . . . . . . . 5 .

3.1 General Description of AFW System . . . . . . 5 3.2 Automatic Initiation. . . . . . . . . . 5 t

i 3.4.1 Evaluation . . . . . . . . . . 5 '

r 3.2.2 Conclusion . . . . . . . . . . 8 l' l 3.3 Flow Indication . . . . . . . . . . . 8 [

3.3.1 Evaluation . . . . . . . . . . 8  ;

3.3. Conclusion . . . . . . . . . . 9 l l 3.4 Descr: tion of Steam Generator Level Indication . . . 9 i l t l

4 CONCLUSIONS . . . . . . . . . . . . . 11' I

5 REFERENCES . . . . . . . . . . . . . 12 i

l 1

l e l  !

l i

s I

i.

4 iii l 0

.ranklin

._ ~Resea_rch .C_ enter

.--v, ,-. .,- n -

i TER-C5257-28 8

1. INTRODUCTION  !

P 1.1 PURPOSE OF REVIEW The purpose of this review is to provide a technical evaluation of the l

emergency feedwater system design to verify that safety-grade automatic .

initiation circuitry and flow indication are provided at North Anna Unit 1.

Although not in the scope of this review, the steam generator level indication available at North Anna Unit 1 is described to assist subsequent NRC staff review.

1.2 GENERIC ISSUE BACKGROUND A post-accident design review by the Nuclear Regulatory Commission (NRC) after the March 28 1979 incident at Three Mile Island (TMI) Unit 2 estab-5 lished that the auxiliary eedwater (AEW) system should be treated as a safety system in a pressurized water reactor (PWR) plant. The designs of safety systems in a nuclear power plant are required to meet general design criteria (GDC) spuified in Appendix A of 10CFR50 [1].

Tt a relevant design er ? teria for the AFW system design are GDC 13, GDC l

20, ant GDC 3 4. goc 13 sets forth the requirement for instrumentation to

! n. nitor variables and systems (over their anticipated ranges of operation) that can affect reactor safety. GDC 20 requires that a protection system be designed to initiate automatically in order to assure that acceptable fuel design limits cre not exceeded as a result of anticipated operational occurrences. GDC 34 requires that the safety function of the designed system, l

that is, the residual heat removal by the AFW system, be accomplished even in the case of a single failure.

! On September 13, 1979, the NBC issued a letter [2] to each PWR licensee j that defined a set of short-term control-grade requirements for the AFW system, specified in NUREG-0578 [3] . It required that the AFW system have automatic initiation and single failure-proof design consistent with the r requirements of GDC 20 and GDC 34. In addition, it required auxiliary feed-water flow indication in the control room in accordance with GDC 13.  !

/%

$ nklin Research Center A Dwmon of The Frenen henatute

, n -

TER-C5257-288 During the week of September- 24, 1979, seminars were held in four regions of the country to discuss the short-term requirements. - On October 30, 1979, another letter was issued to each PWR licensee providing additional clarifica-tion of the NRC staff short-term requirements without altering their intent [4].

Post-TMI analyses of primary system response to feedwater transients and reliability of installed AFW systems also established that, in the long term, the AFW system should be upgraded in accordance with safety-grade requirements.

These long-term requirements were clarified in the letter of 3eptember 5,1980

[5] and formalized in the letter of October 31, 1980 [6]. The October 31 letter incorporated in one document, NUREG-0737 [7], all 'JMI-related items approved ty the commission for implementation.Section II.E.1.2 of NUREG-0737 clarifies the requirements for the AFW system automatic initiation and flow indication.

1.3 PLANT-SPECIFIC BACKGROUND The Licensee of North Anna Unit 1, Virginia Electric and Power Company (VEPCO), provided its initial response to Reference 3 on October 24, 1979

[8]. In this response, VEPCO indicated that the APW system at North Anna Unit 1 was automatically initiated and contained auxiliary feedwater flow indica-tion. Later, further correspondence was issued between VEPCO and the NRC relating to the implementation of NUREG-0578 (9-13]. On December 15, 1980 f i

[14], VEPCO provided its response to the requirements of NUREG-0737. On July j 10, 1981 (15), the NRC forwarded a request for additional information in order f for FRC to complete this review; VEPCO responded to this request on August 3, 1981 [16].  !

f I

t P

{

/h  !

!)00bankhn Research Center A Dmson of The FranWm insatute t

TER-C5257-288  !

2. REVIEW CRITERIA To improve the reliability of the AFW system, the NRC required licensees to upgrade the system, where necessary, to ensure timely automatic initiation .

t when required. The system upgrade was to proceed in two pcases. In the short [

term, as a minimum, control grade signals and circuits were to be used to auto- i matically initiate the AFW system. This control grade system was to meet the l following requirements of NUREG-0578, Section 2.1.7.a [3):  !

"1. The design shall provide for the automatic initiation of r the auxiliary feedwater system.

2. The automatic initiation signals and circuits shall be ,

designed so that a single f ailure will not result in the l loss of auxiliary feedwater system function. )

' t

3. Testability of the initiating signals and circuits shall be a feature of the design.
4. The initiating signals and circuits shall be pcwered from the emergency buses.
5. Manual capability to initiate the auxiliary feedwater sys- i tem from the control room shall be retained and shall be implemented so that a single failure in the manual circuits  ;

will not result in the loss of system function.

6. The cc motor-driven pumps and valves in the auxiliary feed-water system shall be i ncAuded in the automatic actuati n ,

(simultaneous and/or sequential) of the loads to the emer-gency buses.  ;

7. The automatic initiating signals and circuits shall be L designed so that their failure will not result in the loss .

f of manual capability to initiate the AFW system from the Control room." t In the long term, these signals and circuits were to be upgraded in accor-dance with safety-grade requiremen*.s. Specifically, in addition to the above j requirements, the automatic initiation signals and circuits were to have [

iadependent channels, use environmentally qualified components, have system bypassed / inoperable status features, and conform to control system interaction criteria, as stipulated in IEEE Std 279-1971 [17).

/As  !}h nklin Research Center A Dnnaion of The Frannhn insatute

e s

TER-C5257-288 The capability to ascertain the AFW system parformance from the control [

room must also be provided. In the short term, steam generator level f indiertion and flow measurement were.to be used to assist the operator in maintaining the required steam generator level during AFW system operation. I This system was to meet the,following requirements from NUPEG-0578, Section ,

2.1.7.b [3}, as clarified by NUREG-0737,Section II.E..l.2 [7]:  !

r "1. Safety-3rade indication of auxiliary feedwater flow to each steam generator shall be provided in the control room.

2. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency I power . diversity requirements of the auxiliary feedwater system set  !

forth in Auxiliary Systems Branch Technical Position 10-1 of the 3 Standard Review Plan, Section 10. 4. 9 [18] . "

The NRC staff has determined that, in the long term, the overall flowrate .

i indication system for Combustion Engineering and Westinghouse plants should include at least one auxiliary feedwater flowrate indicator and one wide-range i steam generator level indicator for each steam generator or two flow rate indicators. These flow indication systems should be environmentally qualified; powered f rom a highly reliable, battery backed non-class lE power source; periodically testabler part of the plant's QA programs and capable of display on demand.

)

The operator relies on steam generator level instrumentation, in addition to auxiliary feedwater flow indication, to determine AFW sys *.em performance.

The requirements for this steam generator level instrumentation are specified in Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" [19].

l

/k du Franklin Research Center A omma e n n u,n sam.  ;

TER-C5257-288

3. TECHNICAL EVALUATION ,

i 3.1 GENERAL DESCRIPTION OF AFW FYSTEM North Anna Unit 1 is a Westinghouse-designed three-loop nuclear power plant and is essentially identical to North Anna Unit 2. The AFW system supplies water to the secondary side of the steam generators for reactor decay heat removal when normal feedwater sources are unavailable. The AFW system consists of one turbine-driven pump (700 gpm) and two motor-driven pumps (350 gpm each).

Auxiliary feedwater flow to the steam generators is automatically initiated when preset levels of any of several monitored parameters are exceeded. The two motor-driven pumps each discharge through an individual automatic pressure control valve (PCV) into two discharge headers. Either header may supply any steam generator, but both are normally aligned so that each pump supplies a single steam generator. A third header normally provides a flow path from the turbine-driven pump to steam generator A, but can also be cross-connected to either of the AFW headers to supply any steam gener'itor.

Each steam generator is supplied through a normally open, air-operated or motor-operated valve; these valves do not receive any automatic initiation signals. In the event of a pump or piping component failure, the AFW system may be realigned using remote and manually operated valves to ensure flow to at least one good steam generator.

Steam generator level is controlled manually from the control room or the auxiliary shutdown panel by operating the appropriate PCVs in the motor-driven

) pump discharge headers.

3.2 AUTOMATIC INITIATION 3.2.1 Evaluation The AFW system at North Anna Unit 1 is designed as an engineered safe-guards systen to seismic Category I, Class lE, and the automatic initiation signals and ciccuits are designed to comply with the single failure criterion of IEEE Std 279-71 (17].

/w Ubbranklin Research Center A Dnneen of The Frenahn insatute

i TER-C5257-288.

The North Anna Unit 1 AFW automatic initiation system consists of two independent actuation trains. The circuits are powered from emergency ac buses, except the motor-driven pump breaker control and turbine steam admission valves, which are powered from the 125-V de bus. The redundant channels are physically separated and electrically independent. A review of the automatic initiation circuitry revealed no credible single failure that would inhibit the automatic initiation system from groviding AFW flow to at least one good steam generator. The scope of the single failure analysis in this report was limited to the redundancy of power supplies, diversity of actuating signals, and independence and redundancy o* automatic initiation circuits.

The motor-driven pumps are powered by independent ac emergency buses.

The loading of the motor-driven pumps to their respective 4 kV emergency buses is part of the post-accident automatic load sequencing.

The turbine-driven pump receives its steam through either of two parallel air-operated (f ail open) steam admission valves which are supplied from lines that tap off upstream of each stesm generator isolation valve.

The following signals are used for automatic init?.ation of the AFW system motor-driven or turbine-driven pumps:

o low-low steam generator level (2 out of 3 channels on either steam generator) o loss of offsite power o trip of all main feedwater pumps o safety injection.

In addition, the AFW system may be manually initiated from the control room or the auxiliary shutdown panel.

The AFW system and components are tested in accordance with technical specifications. The proper operation of the AFW pumps is checked at least every 31 days, the motor-driven pumps are tested 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to plant heat-up, and the turbine-driven pump is tested 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to plant startup.

Every 18 months each pump is verified to start on receipt of tach of the auto-matic initiation signals. The automatic initiation logic is tested monthly.

/w s b nklin Research Center

~. w ~.~.

'i

?

TER-C5257-288  !

The system design allows one channel to be bypassed for maintenance, -!

testing, and calibration during power operation without initiating a protec- i tive action. Any time a channel is bypassed, the bypass is accompan'ed by a f partial trip alarm and channel status light actuation in the control room.

There are no operating bypasses associated with the AFW system. If some part +

of the system has been administrative 1y bypassed or taken out of service, indication is provided in the control room. .

The only interaction between the AFW system automatic initiation circuits and normal system control functions occurs in the narrow-range steam generator ,

level instrumentation. These level instruments are used for both protection 5

(reactor trip and AFW initiation) and normal control functions (narrow-range channel III only) in the main feedwater system. The control signals are j separated from the protection signals by isolation transformers such that a malfunction in the control circuits will have no ef f ect on the protection signals.

The following individual alarms are provided on the main control board bench section to alert the operator that the AFW equipment may not operate >

properly:

o AFW turbine train A or B not in auto o AFW protection logic in test o AFW pt p pressure control valve not open j o low-low steam generator level (3 channels each)  :

o AFW pump A trip o AFW pump B trip.

The following are the air-operated AFW control valve alarms:

o motor-driven pump supply to steam generator A or B not fully closed o ~ctor-driven pump supply to steam generator C not fully open.

l The following are th' motor-operated AFW control valve alarms:

o Motor-driven pump supply to steam generator A or C not fully closed p Ud nklin Research C ter 4a a n.n uos en ,

e , - n,

i 4

1 I

tar-C5257-2:18 h

o Motor-driven pump supply to steam generator B or turbine-driven (

pump supply to steam generator A not fully open. l A~ review of the automatic and manual initiation circuitry a id signals revealed that no single failure to either circuit train would inhibit the capability for manual initiation from the control room or the auxiliary shutdown panel. The environmental qualification of safety-related elc:trical' and mechanical components, including AFW system circuits and components, is being reviewed separately by the NRC and is not within the FRC scope of review.

3.2.2 Conclusion The initiation signals, logic, and associated circuitry of the automatic initiation feature'of the AFW system of North Anna Unit 1 comply with the long-term safety-grade requirements of NUREG-0578, Section 2.1.7.a, and the subsequent clarification issued by the NRC staff.

3.3 FICW INDICATION 3.3.1 Evaluation The capability to evaluate the performance of the AFW system at North Anna Unit 1 is provided by:

o auxiliary feedwater flow rate (1 channel per steam generator) o narrow-range steam generator level (3 channels each) o wide-range steam generator level (1 channel each) o AFW pump discharge and suction pressure o AFW pump status indicators o discharge valve position indicators o pump current and voltage 4

o condensate storage tank level.

The Licensee has stated that the AFW flow indication for each steam generator is safety grade. The individual flow indication circuitry is powered from separate vital buses. AFW flow indication at North Anna Unit I gh UOO nklin Research Center A Desu3n of The Franabn insatute

TER-C5257-288 is not designed to accommodate a single failure; however, safety-grade wide-range steam gererator level indication is provided as a backup. The AFW flow indicators are testable from the transmitter to the indicator. The overall accuracy of the feed flow loops is within +10%. The flow indication channels are tested in accordance with technical specifications.

The environmental qualification of the AFW flow indicators will be rarvirwed separately by the NRC and is not within the FRC scope of review.

3.3,2 Conclusion It is concluded that the auf 2ary feedwater flow instrumentation at North Anna Unit 1 complies with the long-term safety-grade requirements of NUREG-0578, Section 2.1.7.b, and the subsequent clarification issued by the NaC.

3.4 DESCRIPTION

OF STEAM GENERATOR LEVEL INDICATION Steam generator level indication at North Anna Unit I consists of three safety-grade narrow range level channels and one safety-grade wide-range level channel per steam generator. The level transmitters and their power supplies are as follows:

STEAM GENERATOR A Channel Transmitter Vital Bus Wide Range LT-1477 1-II Narrow Range I LT-1474 1-I Narrow Range II LT-1475 1-II Narrow Range III LT-1476 1-III STEAM GENERATOR B Channel Transmitter Vital Bus Wide Range LT-14 8 7 1-III Narrow Range I LT-1484 1-I Narrow Range II LT-148 5 1-II Narrow Range III LT-1486 1-III nklin Research A Dhnseon of The Frankha

  • I

-i

. TER-C5257-288 t i

STEAM GENERATOR C

.I Channel Transmitter Vital Bun Wide Range LT-1497 1-IV ,

Narrow Range I LT-1494 1-1 i Narrow Range II LT-1495 1-II l Narrow Range III LT-1496 1-III ,

I Calibration and testing is performed once every 18 months for l narrow-range level channels and once every 24 months for wide-range level

{

channels.  :

i

'Ibe wise-range channels for all three steam generators are indicated individuall) on one stripchart recorder. Narrow-range channels I and II for

,i all three st eam generators are indicated on vertical gages. Narrow-range  ;

channel III for all.three steam generators are indicated on both a vertical  ;

gage and a stripchart recorder. f i

l I

i f

P r

6 1

t s

i gg t

' ud0 Franklin Researth Center A Dmeson of The Fronthn anssture

2 l

t TER-C5257-286, i

4. CONCLUSIONS l

I The initiation signals, logic, and associated circuitry of the North Anna Unit 1 auxiliary feedwater system comply with the long-term safety-grade  ;

i res,mirements of NUREG-0578, Section 2.1.7.a [3), and the subsequent clarifi-  ;

cation 1*. sued by the NRC. '

I The auxiliary feedwater flow instr wentation at North Anna Unit 1 complies with the long-term safety-grade requirements of NUREG-0578, Section 2.1.7.b [3], and the subsequent clarification issued by the NRC. j I

i r

/A dbd aFranklin Research Center o, on or w r,.non wou.

+

l i

s TER-C5257-2BS F. REFERENCES f 7

1. Code of Federal Regulations, Title 10, Office of the Federal l Register, National Archives and Records Service, General Services 1 Administration, Revised January 1,1980.

t

2. NRC, Generic letter to all PWR licensees regarding short-term .

requirements resulting from Three Mile Island Accident, September  ;

13, 1979. e

3. NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and i

Short-Term Recommendations," USNRC, July 1979.

4. NBC, Generic letter to all PWR licensees clarifying lessons learned [

short-term requirements, October 30, 1979. j t

5. NBC, Generic letter to all PWR licensees regarding short-term

requirement resulting from Three Mile Island Accident, Eeptember 5, 1980.

6. NBC Generic letter to all PWR licensees regarding post-TMI I

requirements, October 31, 1980.

7. FUREG-0737, " Clarification of TMI Action Plan Requirements," USNRC, November 1980.
8. C. M. Stallings (VEPCO) ,

Letter to H. R. Denton (NRC) [

October 24, 1979

9. C. M. Stallings (VEPCO)

Letter to H. R. Denton (NRC) ,

November 26, 1979

10. C. M. Stallings (VEPCO)

Letter to H. R. Denton (NRC)

December 28, 1979.

11. C. M. Stallings (VEPCO)

Letter to H. R. Denton (NRC)

I January 10, 1980.

12. C. M. Stallings (VEPCO)  ;

Letter to H. R. Denton (NRC)

February 25, 1980.

13. B. R. Sylvia (VEPCO)

Letter to H. R. Denton (NRC)

July 7,1980. l

_nklin_Resea_rch

_ . C_ enter

?

TER-C5257-288-

14. B. R. Sylvia (VEPCO)

Letter to H. R. Denton (NRC)

December 15, 1980

15. R. A. Clark (NRC)

Letter to R. H. Leasburg (VEPCO)

July 10, 1981

16. R. H. Leasburg (VEPCO)

Letter to R. A. Clark (NRC)

August 3, 1981

17. IEEE Std 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stat!ons," Institute of Electrical and Electronics Eng inee rs, Inc., New York, NY.
18. NUREG-75/087, " Standard Review Plan," Section 10.4.9, Rev. 1, USNRC, no date.
19. Regulatory Guide 1.97 (Task RS 917-4), " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Rev. 2, NRC, Decembec 1980.

t

?

i.

+

s 1

i 1

1 N

P l

_nklin_Resea_rch

_ . Ce_nter .

I